On the options for incorporating nuclear data uncertainties in criticality safety assessments for LWR fuel

Vasiliev, A.; Rochman, D.; Pecchia, M.; Ferroukhi, H.

doi:10.1016/j.anucene.2018.01.046

Cited by 11 publications

(11 citation statements)

References 25 publications

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“…The application of a best-estimate computational route is now complemented with conservative coverage in the form of a stochastic analysis of the main uncertainty components in combination with the application of bounding burnup profiles, as illustrated below. Furthermore, the USL value considered in this final stage of the work is now based on the comprehensive validation study for LWR fuel performed at PSI for the MCNP code in combination with the ENDF/B-VII.1 library, which has been selected for the routine criticality calculations [10,11].…”

Section: Methodology For Preliminary Loading Curve Derivationmentioning

confidence: 99%

“…The KeffLTB|AOA term stands for the Lower Tolerance Bound for the particular Area of Applicability (AOA; here, this is limited to Swiss LWR fuel), and its value was reported in Reference [10] as 0.99339 (following the Gaussian-based derivation; see Reference [10] for details) for the PSI CSE methodology using the MCNP code in conjunction with the ENDF/B-VII.1 library. As outlined in Section 2.4, ΔkeffAM is the “administrative margin” normally imposed to cover the unknown uncertainties to ensure subcriticality, which is assumed here to be 0.05000 (The administrative margin to criticality is normally set at 5000 pcm, i.e., the k eff of the system plus the calculation bias and uncertainty in the bias should not exceed 0.95.…”

Section: Methodology For Preliminary Loading Curve Derivationmentioning

confidence: 99%

“…Therefore, it will be conservative to assume a full correlation between both components and thus to estimate the total ND-related keff component according to Equation (3). More advanced assessments are planned for the near future with the latest release of the ENDF/B library (ENDF/B-VIII), including the analysis of the ND-related correlations, as proposed in References [10,29].…”

Section: Quantification Of the Bounding Effects And Random Uncertamentioning

confidence: 99%

“…In particular, the validation studies for the fuel depletion calculations with the stand-alone CASMO code [6,7] and with the CASMO/SIMULATE codes using PSI proprietary post-irradiation examination (PIE) data [8], for the reactor core-follow simulations with the CASMO/SIMULATE codes using the in-core reactor measurements [9] and for the criticality calculations with the Monte Carlo N-Particle ® MCNP ® or MCNPX Software (The registered trademarks are owned by Triad National Security, LLC, manager and operator of Los Alamos National Laboratory (see ; assessed on 29 January 2019).) using criticality benchmark experiments [10,11] should be mentioned as examples.…”

Section: Introductionmentioning

confidence: 99%

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Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

et al. 2019

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This paper presents preliminary criticality safety assessments performed by the Paul Scherrer Institute (PSI) in cooperation with the Swiss National Cooperative for the Disposal of Radioactive Waste (Nagra) for spent nuclear fuel disposal canisters loaded with Swiss Pressurized Water Reactor (PWR) UO2 spent fuel assemblies. The burnup credit application is examined with respect to both existing concepts: taking into account actinides only and taking into account actinides plus fission products. The criticality safety calculations are integrated with uncertainty quantifications that are as detailed as possible, accounting for the uncertainties in the nuclear data used, fuel assembly and disposal canister design parameters and operating conditions, as well as the radiation-induced changes in the fuel assembly geometry. Furthermore, the most penalising axial and radial burnup profiles and the most reactive fuel loading configuration for the canisters were taken into account accordingly. The results of the study are presented with the help of loading curves showing what minimum average fuel assembly burnup is required for the given initial fuel enrichment of fresh fuel assemblies to ensure that the effective neutron multiplication factor, keff, of the canister would comply with the imposed criticality safety criterion.

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Section: Methodology For Preliminary Loading Curve Derivationmentioning

confidence: 99%

Section: Methodology For Preliminary Loading Curve Derivationmentioning

confidence: 99%

Section: Quantification Of the Bounding Effects And Random Uncertamentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

et al. 2019

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“…PCC computed in correlation studies can be interpreted as the measure of similarity between the tested systems, or, similarly, as a representativity measure of a benchmark model with respect to a certain application system. Up to now, the correlation studies with NUSS have covered both Criticality Safety and Dosimetry applications, as can be found in [15], [16], [17].…”

Section: Nuss: Nuclear Data Stochastic Sampling With Ace Filesmentioning

confidence: 99%

Assessment of representativity of the PETALE experiments for validation of Swiss LWRs ex-core dosimetry calculations

et al. 2020

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The international experimental program PETALE will be carried out at the CROCUS research reactor of EPFL. The program aims at measuring neutron penetration in slabs made of materials composing typical LWR reactor pressure vessel. The measurements will be used for code and nuclear data validation and for the evaluation of the reflecting properties of these materials. In this paper the representativity of the PETALE experiments is assessed with respect to operational LWR reactors dosimetry and activation evaluations using the Paul Scherrer Institute (PSI) in-house tool NUSS. The NUSS tool allows the stochastic sampling of nuclear data using covariance matrices available in modern nuclear data libraries and the subsequent running of a Monte Carlo code with the modified data files. The representativity can then be assessed based on the Pearson correlation coefficients. The ultimate goal of the work is first of all to assess if the planned PETALE measurements could be applicable beyond their primary purpose and serve for extending the PSI validation database for LWR reactor dosimetry evaluations. Secondly, provided that the PETALE measurements are found useful for the task above, the information on the correlations between the PETALE neutron detectors’ responses and the reactor dosimetry quantities of interest, shall be presented and discussed.

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Study on strip-shaped liquid film in the corrugated plate dryer

Wang

Chen

et al. 2020

Annals of Nuclear Energy

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On the options for incorporating nuclear data uncertainties in criticality safety assessments for LWR fuel

Cited by 11 publications

References 25 publications

Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

Preliminary Assessment of Criticality Safety Constraints for Swiss Spent Nuclear Fuel Loading in Disposal Canisters

Assessment of representativity of the PETALE experiments for validation of Swiss LWRs ex-core dosimetry calculations

Study on strip-shaped liquid film in the corrugated plate dryer

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